• Corrosion Science and Technology
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• v.11 no.6
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• pp.225-231
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• 2012

Fitness-for-Service is a useful technology to determine replacement timing, next inspection timing or in-service when nuclear power plant's buried pipes are damaged. If is possible for buried pipes to be aged by material loss, cracks and occlusion as operating time goes by. Therefore Fitness-for-Service technology for buried pipe is useful for plant industry to perform replacement and repair. Fitness-for-Service for buried pipe is studied in terms of existing code and standard for Fitness-for-Service and a current developing code case. Fitness-for-Service for buried pipe was performed according to Code Case N-806 developed by ASME (American Society of Mechanical Engineers).

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.17 no.1
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• pp.8-17
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• 2021

The seat leakage test is required after the maintenance work on the valve seat. Either the test has been performed outside of the plant after cutting the valve from the pipe system or the simplified test has been performed so far. It was unable to perform the test at the plant site because it is hard to make a steady pressure on the valve inlet when it is installed in the pipe. This research aims to perform the leakage test in the nuclear power plant while it is installed in the pipe system. The mock-up test is performed by pressurizing the leak-off pipe on the valve body. The result is compared with traditional test result by pressurizing the valve inlet. Furthermore the chamber mock-up tests are performed under various conditions. The leak rate by the developed test using the leak-off pipe is found to be similar but greater than the leak rate by the existing test method. It implies that the test using the leak-off pipe is more conservative than the existing test. The methodology and the equipment which this paper suggests that on-site seat test is possible and the application of the technology could reduce the time and cost for the valve maintenance work significantly.

• The Journal of Korea Robotics Society
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• v.17 no.2
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• pp.159-163
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• 2022

A linearly moving structure in the area where the friction force is dominant - such as ducts filled with grease in the nuclear power plant - experiences increase in friction since the contact surface gets larger as the structure proceeds. To solve this problem is critical for the pipe inspection robot to investigate further area and this makes the system more energy-efficient. In this paper, we propose a passively growing sheath that can be added to linearly moving structures using zipper mechanism. The mechanism enables the linearly moving structures to maintain rolling contact condition against external environment, which provides substantial reduction in kinetic friction. To analyze the effect of the mechanism's head shape, we establish a physical model and compare to the experimental results. Finally, we have shown that the passively growing sheath can be successfully applied to the pipe inspection robot for the nuclear power plant.

• Nuclear Engineering and Technology
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• v.30 no.2
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• pp.164-172
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• 1998

This paper presents the detailed modelling of reactor vessel ledge region for the dynamic analysis of the coupled internals and core model. The dynamic responses due to earthquake and pipe break are calculated using the input motions of reactor vessel taken from Ulchin nuclear power plant units 3 and 4. Two different representations for detailed and simplified models of the RV ledge region are made. The dynamic responses of the reactor internals components are compared between them. Response characteristics are reported and simplified model is suggested for earthquake and pipe break analysis for the future design of the reactor internals.

• Journal of the Korea institute for structural maintenance and inspection
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• v.23 no.6
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• pp.77-83
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• 2019

Upon installing a seismic isolation device on a nuclear power plant, the device takes on the suppression of seismic loads. This is expected to bring about a larger displacement than what is seen prior to the installation of the seismic isolation device. Depending on the displacement change, the seismic risk for some equipment can increase. Particularly in case of the piping system, which is used for connecting the structure isolated from seismic events with common structures, the seismic risk is expected to rise significantly. In this study, the limit state of the steel pipe tee, which is a vulnerability part of the nuclear power plant piping system, was defined as leakage, and an in-plane cyclic loading test was conducted. As it is difficult to measure the moment and rotation of the steel pipe tee using the conventional sensors, an image signal was used. This study proposed a leakage line and low-cycle fatigue curves using the relationship between the moment and the rotation of a 3-inch steel pipe tee.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.15 no.8 s.101
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• pp.988-995
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• 2005

Vibration has been severly increased at the branch pipe of main steam header since the commercial operation of nuclear power plant. Intense broad band disturbance flow at the discontinuous region such as elbow, valve, and header generates the acoustical pulsation which is propagated through the piping system. The pulsation becomes the source of low frequency vibration at piping system. If it coincide with natural frequency of the pipe system, excessive vibration is made. High level vibration due to the pressure pulsation related to high dynamic stress, and ultimately, to failure probability affects fatally the reliability and confidence of plant piping system. This paper discusses vibration effect for the branch pipe system due to acoustical pulsations by broad band disturbance flow at the large main steam header in 700 MW nuclear power plant. The exciting sources and response of the piping system are investigated by using on-site measurements and analytical approaches. It is identified that excessive vibration is caused by acoustical pulsations of 1.3 Hz, 4.4 Hz and 6.6 Hz transmitted from main steam balance header, which are coincided with fundamental natural frequencies of the piping structure. The energy absorbing restraints with additional stiffness and damping factor were installed to reduce excessive vibration.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2005.05a
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• pp.780-785
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• 2005

Vibration has been severly increased at the branch pipe of main steam header since the commercial operation of a nuclear power plant. Intense broad band disturbance flow at the discontinuous region such as elbow, valve or heather generates the acoustical pulsation which is propagated through the piping system. The pulsation becomes the source of low frequency vibration at piping system. If it coincide with natural frequency of the pipe system, excessive vibration is made. High level vibration due to the pressure pulsation related to high dynamic stress, and ultimately, to failure probability affects fatally the reliability and confidence of plant piping system. This paper discusses vibration effect for the branch pipe system due to acoustical pulsations by broad band disturbance flow at the large main steam header in 7nn nuclear power plant. The exciting sources and response or the piping system are investigated by using on site measurements and analytical approaches. It is identified that excessive vibration is caused by acoustical pulsations of 1.3Hz, 4.4Hz and 6.6Hz transferred from main steam header, which are coincided with fundamental natural frequencies of the piping structure. The energy absorbing restraints with additional stiffness were installed to reduce excessive vibration.

• Journal of the Korean Society for Nondestructive Testing
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• v.33 no.4
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• pp.323-329
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• 2013

This perform research of angle rated defect detection conditions and nuclear power plant piping defect detection by lock-In infrared thermography technique. Defects were processed according to change for wall-thinning length, Circumference orientation angle and wall-thinning depth. In the used equipment IR camera and two halogen lamps, whose full power capacitany is 1 kW, halogen lamps and target pipe's distance fixed 2 m. To analysis of the experimental results ensure for the temperature distribution data, by this data measure for defect length. Reliability of lock-In infrared thermography data is higher than Infrared thermography data. This through research, Shape of angle rated defect is identified industry place. It help various angles defect detection in the nuclear power plant in operation.

• The KSFM Journal of Fluid Machinery
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• v.6 no.4 s.21
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• pp.45-49
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• 2003

This work focuses on eliminating tiny particles from the coolant in a nuclear pipe line by using a permanent magnet on the exterior surface of the pipe. This method have some merits compared with the currently applied methods and is expected to be applied to most of the pipe lines in the nuclear plant. For instance in this method, a ring is attacked to the exterior surface of the pipe, so that it does not affect the inflows directly. Further, the cost needed in the initial build-up of the facility is low.

• Nuclear Engineering and Technology
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• v.39 no.3
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• pp.215-220
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• 2007

The existing flooding Probabilistic Safety Analysis(PSA) was updated to reflect the Korean plant specific operating experience data into the flooding frequency to improve the PSA quality. Both the Nuclear Power Experience(NPE) database and the Korea Nuclear Pipe Failure Database(NuPIPE) databases were used in this study, and from these databases, only the Pressurized Water Reactor(PWR) data were used for the flooding frequencies of the flooding areas in the primary auxiliary building. With these databases and a Bayesian method, the flooding frequencies for the flooding areas were estimated. Subsequently, the Core Damage Frequency(CDF) for the flooding PSA of the Ulchin(UCN) unit 3 and 4 plants based on the Korean Standard Nuclear Power Plant(KSNP) internal full-power PSA model was recalculated. The evaluation results showed that sixteen flooding events are potentially significant according to the screening criterion, while there were two flooding events exceeding the screening criterion of the existing UCN 3 and 4 flooding PSA. The result was compared with two kinds of cases: (1) the flooding frequency and CDF from the method of the existing flooding PSA with the PWR and Boiled Water Reactor(BWR) data of the NPE database and the Maximum Likelihood Estimate(MLE) method and (2) the flooding frequency and CDF with the NPE database(PWR and BWR data), NuPIPE database, and a Bayesian method. From the comparison, a difference in CDF results was revealed more clearly between the CDF from this study and case (2) than between case (1) and case (2). That is, the number of flooding events exceeding the screen criterion further increased when only the PWR data were used for the primary auxiliary building than when the Korean specific data were used.